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Vver and Rbmk Cross Section Libraries for Origen-Arp

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Vver and Rbmk Cross Section Libraries for Origen-Arp VVER AND RBMK CROSS SECTION LIBRARIES FOR ORIGEN-ARP Germina Ilas, Brian D. Murphy, and Ian C. Gauld, Oak Ridge National Laboratory, USA Introduction An accurate treatment of neutron transport and depletion in modern fuel assemblies characterized by heterogeneous, complex designs, such as the VVER or RBMK assembly configurations, requires the use of advanced computational tools capable of simulating multi-dimensional geometries. The depletion module TRITON [1], which is part of the SCALE code system [2] that was developed and is maintained at the Oak Ridge National Laboratory (ORNL), allows the depletion simulation of two- or three-dimensional assembly configurations and the generation of burnup-dependent cross section libraries. These libraries can be saved for subsequent use with the ORIGEN-ARP module in SCALE. This later module is a faster alternative to TRITON for fuel depletion, decay, and source term analyses at an accuracy level comparable to that of a direct TRITON simulation. This paper summarizes the methodology used to generate cross section libraries for VVER and RBMK assembly configurations that can be employed in ORIGEN-ARP depletion and decay simulations. It briefly describes the computational tools and provides details of the steps involved. Results of validation studies for some of the libraries, which were performed using isotopic assay measurement data for spent fuel, are provided and discussed. Cross section libraries for ORIGEN-ARP Methodology The TRITON capability to perform depletion simulations for two-dimensional (2-D) configurations was implemented by coupling of the 2-D transport code NEWT with the point depletion and decay code ORIGEN-S. NEWT solves the transport equation on a 2-D arbitrary geometry grid by using an SN approach, with a treatment of the spatial variable that is based on an extended step characteristic method [3]. The flux solution from the transport calculation serves for updating, at each depletion step, the effective cross sections used by ORIGEN-S in the depletion calculation. The nuclide composition provided by ORIGEN-S is applied to a subsequent transport calculation, and the process continues in an iterative manner until all the depletion steps are simulated. At the end of each NEWT transport calculation, weighted cross sections and fluxes are obtained, which are subsequently used in the COUPLE code to calculate one-group effective cross sections required in the ORIGEN-S calculations. More details on the components, functionality, and capabilities of TRITON can be found in Ref. [4]. The cross sections used in the NEWT transport calculation in this work were from a SCALE 44- group ENDF/B-V transport library; the BONAMI and NITAWL modules were used to correct the cross sections for temperature and resonance effects. Note that the current default option for resonance self-shielding treatment in SCALE 5.1 is CENTRM/PMC. CENTRM is a code that solves the transport equation in a one-dimensional (1-D) cell by using a combination of point-wise and multigroup cross section data, and provides a quasi-continuous problem-dependent energy flux spectrum; this flux is used by the PMC module to generate self-shielded multigroup cross sections for the transport calculation. However, the CENTRM/PMC methodology was not used for the work reported in this paper due to limited experience with its use for reactor operating conditions. The ORIGEN-ARP sequence in SCALE includes a Windows graphical user interface and permits the user to create and execute depletion and decay cases with minimal effort, and to generate tables and plots showing the calculated data in a fast and user-friendly fashion. ORIGEN-ARP has three main components: (1) the ARP code to interpolate on a set of pre-generated burnup- dependent cross sections to obtain cross sections for use with ORIGEN-S; (2) the ORIGEN-S code to perform depletion and decay simulations; and (3) the OPUS/PlotOPUS codes to extract and plot 2 the calculated results. The interpolation in ARP, which is based on Lagrange polynomials, can be carried out in the case of a uranium dioxide fuel for three interpolation parameters: burnup, fuel enrichment, and moderator density. The main computational steps used in this work to obtain the cross sections for use with ORIGEN- ARP consisted of the following. (1) For each unique assembly configuration, and for combinations of discrete values of enrichment and moderator density, TRITON depletion simulations were carried out for a given number of depletion steps to generate burnup-dependent cross sections for use with ORIGEN-S. (2) The number of cross sections sets in each library file was reduced by using the ARPLIB utility code in SCALE, in order to decrease the library size; the thinning of a library file was performed by removing the cross section data for those burnup points in the range where the cross sections exhibit a slow variation with burnup. VVER assemblies The generation of the VVER libraries was carried out to provide data for modern assembly designs that are currently used in the nuclear industry. The cross section libraries, which are briefly described in Table 1, were created for five representative VVER assembly configurations—one VVER-1000 and four VVER-440 configurations—and cover a burnup range up to 70.5 GWd/MTU. The burnup values corresponding to data in each library are 0, 1.5, 4.5, 7.5, 10.5, 13.5, 16.5, 31.5, 46.5, 58.5, and 70.5 GWd/MTU. Three of the four VVER-440 configurations are characterized by an enrichment zoning with an assembly average enrichment of 3.82, 4.25, and 4.38 wt % 235U; two of these designs contain burnable absorber rods. The other VVER-440 design considered is characterized by a flat enrichment; in this case, libraries were generated for three enrichment values: 1.6, 2.4, and 3.6 wt % 235U. The VVER-1000 libraries were generated for six fuel enrichment values: 1.5, 2.0, 235 3 3.0, 4.0, 5.0, and 6.0 wt % U. A core-average value of 0.75 g/cm was used for the moderator density in all cases. The level of detail in the TRITON models is illustrated in Figure 1 for a VVER- 440 assembly design with 3.82 wt % 235U average enrichment and also for a VVER-1000 design. The detailed assembly data and reactor operating data used to generate the libraries can be found in the SCALE 5.1 manual. The VVER libraries previously released with SCALE 5.0 were generated using the 1-D depletion sequence of SAS2. The former VVER-440 libraries were shown in a previous study to provide sufficiently accurate results for actinides as compared to measured isotopic assay data [5]. Table 1. Description of the VVER libraries for ORIGEN-ARP in SCALE 5.1 Assembly design Name of library Enrichment profile Enrichment value(s) (wt % 235U) VVER-440 vver440(3.6) flat 1.6, 2.4, 3.6 VVER-440 vver440(3.82) zoned average 3.82 VVER-440 vver440(4.25) zoned average 4.25 VVER-440 vver440(4.38) zoned average 4.38 VVER-1000 vver1000 flat 1.5, 2.0, 3.0, 4.0, 5.0, 6.0 3 (a) VVER-440 (3.82) (b) VVER-1000 Figure 1. TRITON models for VVER assemblies. RBMK assembly The RBMK libraries were generated based on design data representative of the Chernobyl Unit 4 reactor. They were obtained by using a TRITON model as illustrated in Figure 2 that shows a single fuel assembly and surrounding graphite moderator. The assembly fuel rods are surrounded by coolant water within a zirconium cylindrical container in the graphite moderator. The coolant water also acts as a moderator. The RBMK libraries developed are representative of the average assembly. For illustrative purposes, the model in Figure 2 distinguishes between the inner and outer rings of fuel rods to illustrate that separate libraries representative of either the inner or outer fuel can be created in addition to those representative of the average assembly. Cross sections for a burnup range up to approximately 25 GWd/MTU were generated for each combination of values from a set of fuel enrichments and coolant densities. In contrast to a VVER reactor, for which the use of an average coolant density is sufficiently adequate given its small variation axially, the RBMK configuration has a boiling water coolant and therefore an axial void- fraction profile. Libraries were prepared for three enrichments (1.8, 2.0, and 2.2 wt % 235U) and six coolant densities (0.15, 0.28, 0.41, 0.54, 0.67, and 0.80 g/cm3). The libraries were provided for fresh fuel (zero burnup) and for twenty other burnup values between 0.625 and 24.375 GWd/MTU with a burnup step of 1.25 GWd/MTU. Figure 2. TRITON model for RBMK assembly. 4 Validation studies Validation studies of the libraries were performed by comparison to available experimental data from selected isotopic assay measurements for VVER and RBMK spent fuel. VVER libraries Validation studies for the VVER libraries were carried out by using isotopic assay measurement data for spent fuel from experiments performed at the Khlopin Radium Institute (KRI) in Russia [6,7,8]. The measured samples included in the present analysis were selected from VVER-440 and VVER-1000 fuel assemblies that were irradiated in reactors operated at the Novo-Voronezh and Kalinin nuclear power plants. The initial fuel enrichment was 3.6 wt % 235U and 4.4 wt % 235U for the samples from VVER-440 and VVER-1000 fuel, respectively. In both cases, a flat enrichment was modeled. The analysis in this work included twenty VVER-440 fuel samples with burnups in the range 20–43 GWd/MTU and thirteen samples from VVER-1000 fuel with burnups between 14 and 52 GWd/MTU.
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